1. Field of the Invention
The invention relates to a pressurized-water-cooled fuel assembly and to a fuel rod of such a fuel assembly having a cladding tube that encloses a fuel filling and which includes a first thicker inner layer that faces toward the fuel filling and is composed of a first zirconium alloy, and a second thinner outer layer which is metallurgically bound to the inner layer and is composed of a second zirconium alloy, the two zirconium alloys each containing at least the metals tin, iron and chromium as alloying constituents.
Zirconium is a comparatively soft metal which is particularly suitable for structural parts of nuclear reactors because of its low neutron absorption. It is technically generally manufactured as "sponge zirconium" having maximum impurities which are standardized for use in nuclear reactors. Since the fuel rods of reactor fuel assemblies are only finger-thin but are several meters long, a high strength which is constant even after prolonged irradiation and which is achieved by adding tin to the alloy is necessary for the fuel rods that are filled with fuel and for the guide tubes, spacers and other structural parts of the fuel assembly. In water, pure zirconium forms a thin oxide layer which protects the metal from further oxidation, but at the same time impurities incorporated in the structure of the zirconium or the oxide layer, particularly nitrogen, may considerably accelerate corrosion. Although the tin addition neutralizes the corrosive action of nitrogen, particularly in conjunction with small additions of iron, it promotes the corrosion itself at higher tin contents (as explained in the publication by B. Lustman and Frank Kerze entitled: "The Metallurgy of Zirconium", New York, 1955, page 538, FIG. 10.34 and page 628, FIG. 11.35). The iron addition hardens the alloy so that, even at comparatively low iron contents, processing the alloy to produce thin, long cladding tubes is virtually no longer possible. In addition, such an iron content results in an increased diffusion and absorption in the metal of hydrogen produced during the corrosion in water, with hydrogenated regions being formed (as explained in the publication by W. Berry, D. Vaughan and E. White entitled: "Hydrogen Pickup During Aqueous Corrosion of Zirconium Alloys" in: Corrosion, Vol. 17, No. 3, March 1961, page 109 t, FIG. 1) which are very brittle and drastically reduce the mechanical robustness of the material.
Therefore, on the basis of corrosion experiments in the laboratory in which the temperature was raised to accelerate the corrosion experiments, an optimum range was specified for the contents of tin and iron, which range was optimized as far as possible with regard to corrosion by additionally taking chromium and nickel alloying constituents into account, while attention was also being paid at the same time to an adequate mechanical and thermal robustness. The robustness investigations were partly carried out under reactor conditions in order to obtain adequate strengths even after prolonged irradiation. In the meantime, the alloys "Zircaloy 2" and "Zircaloy 4" developed in that process have proved essentially satisfactory in structural parts of fuel assemblies of water-cooled reactors. Table 1 shows the grades of sponge zirconium, Zircaloy 2 and Zircaloy 4 which are permitted as materials in reactor technology. In that publication, the amounts of the alloying constituents are specified as percentages by weight, based on the alloy.
Upon prolonged use in the reactor core, the fuel releases iodine and other fission products, with the result that an aggressive atmosphere with gradually increasing pressure builds up therein and the volume of the fuel itself also increases. Particular mechanical-thermal-chemical stresses which may result in the destruction of the cladding tube and make it necessary to interrupt the reactor operation to replace the fuel rod therefore occur on the inside of the cladding tube. In relation to boiling-water reactors in particular, use is therefore frequently made of composite tubes which have, on their inside, a thin layer of pure zirconium or of an alloy which is optimized in relation to higher ductility and a resistance which is adapted to the chemical, mechanical and thermal conditions. The remaining, thick outer layer of the cladding tube on one hand provides for the necessary mechanical robustness of the entire cladding tube and on the other hand is proof against corrosion under the conditions of the coolant. A liquid/vapor mixture mainly includes water at moderate pressure and moderate temperature, and includes the Zircaloy 2 already mentioned.
Modern pressurized-water reactors have a fuel assembly structure and reactivity distribution which make it possible to increase the cost effectiveness as a result of long service lives, high burn-ups and increased operating temperatures. However, a requirement in that connection is that the fuel assembly failure probability due to cladding tube defects is kept extremely low. In that connection, the emphasis is on efforts to eliminate primary damage to the outer surface of the cladding tube since, as a result of such primary defects at any point in the long cladding tube, the high pressure in the coolant flow can force water into the cladding-tube interior where it reacts with the hot fuel. The consequence may then be devastating secondary damage on the inside of the cladding tube.
Therefore, for the first time, Published European Application No. 0 212 351 A1, corresponding to U.S. Pat. No. 4,735,768, proposes, as a cladding tube for a water-cooled reactor fuel assembly, a double-layer composite tube having an inner layer, adjacent the fuel, which occupies 80 to 95% of the total wall thickness of the cladding tube and is formed of Zircaloy 4, while the thin outer layer is formed of zirconium containing 0.5% iron and 0.25% vanadium. The so-called "duplex tube", which is used as the carrier that determines the mechanical properties of the entire cladding tube, generally has a thick layer of Zircaloy 2 or Zircaloy 4 to which a thin, outer protective layer of a second zirconium alloy is metallurgically bonded (for example by combined extrusion of two concentric tubes). The alloy contains 0.1 to 1% vanadium and/or 0.1 to 1% platinum and/or 1 to 3% copper, and optionally up to 1% iron. All of the percentages are based on the weight of the alloys.
Such cladding tubes exhibit an excellent behavior and, in particular, even at burn-ups of between 40 and 60 MWd/kg of uranium, only oxide layers having thicknesses below 20 .mu.m occur at the outer surface which is exposed to the pressurized water, while the mechanical behavior, such as the increase in length and the shrinkage of the fuel-rod diameter, is within the range of the most favorable values obtained with one-piece tubes of Zircaloy 4. However, those cladding tubes are comparatively expensive, and in particular, the outer alloy is difficult to process mechanically because of its hardness and requires lengthy, careful processing steps with an increased reject occurrence. In addition, the alloying constituents of the second alloy are not permitted per se as reactor materials because of the high neutron absorption and can only be tolerated because they are used only in the thin outer layer in low concentrations. However, the reject material produced during the cladding-tube manufacture cannot readily be fed back into the cladding-tube production since the alloying constituents contained in the second alloy result in impurities in the zirconium or zircaloy which cannot be tolerated. The excellent results of that duplex cladding tube is attributed, in particular, to the absence of tin in the outer layer.
Published European Application No. 0 301 295 A1, corresponding to U.S. Pat. No. 4,963,316, describes a duplex cladding tube which is easier to process and in which the outer layer is formed of a tin-free zirconium alloy containing 2.5% niobium or at least a low-tin alloy containing 0.25% tin, 0.5% iron and 0.05% chromium. Good results are expected for a niobium content between 0.2 and 3% and/or a total content of iron, chromium, nickel and tin of between 0.4 and 1.0%. U.S. Pat. No. 5,023,048 describes a similar fuel rod in which the outer layer is formed of Zr, (0.35 . . . 0.65) % Sn, (0.2 . . . 0.65) % Fe, (0.24 . . . 0.35) % Nb and (0.09 . . . 0.16) % O and contains no chromium.
Whereas the fuel assemblies of boiling-water reactors are exposed to a coolant temperature of about 280.degree. C. at a pressure of 70 bar, the surface temperature of the cladding tubes of pressurized-water reactors is about 340.degree. C., with the coolant having an outlet temperature of about 320.degree. C. at 170 bar. Despite the fact that the differences in the operating conditions at first appear relatively small, the corrosion processes occurring under those circumstances differ markedly. Laboratory experiments which were carried out at elevated temperatures (for example 360.degree. to 550.degree. C.), that is to say in the vicinity of or above the critical temperature of water, in order to shorten the test times and to intensify the corrosion conditions in the development of Zircaloy 2 and Zircaloy 4, are therefore only of limited meaningfulness for the corrosion behavior.
In addition, measures which may have an effect on the corrosion processes occurring at the fuel rods are taken by individual power-station operators for other reasons, for example to prevent corrosion of heat exchangers or other components in the coolant circuit. Measures which may result in an alteration in the corrosion occurring at the fuel-rod claddings are also taken during start-up or during certain temporary operating states.
Thus, for example, fuel rods of pressurized-water reactors which are operated at a coolant outlet temperature of 316.degree. C. on average may exhibit an excellent behavior but may exhibit a higher failure probability even in reactors having mean coolant outlet temperatures of 326.degree. C. which, although slight, is undesirable. One of the causes thereof may be the strong temperature dependence of hydrogen diffusion during very long service lives. At the start of the operation of some nuclear reactors, a small amount of dissolved lithium hydroxide, which may considerably affect the corrosion of the cladding tubes, is present in the coolant circuit.
In particular, an increase in fuel-rod power may result in a fuel-rod outer temperature which is in fact only a few degrees higher but at which local boiling resulting in substantially more severe corrosion conditions occurs in the pores of the oxide layer. However, the Li content in the cooling water, which is not enough to cause problems in conventional fuel-assembly structures, may concentrate in the pores during local boiling and make it necessary either to change over to other alloys or to dispense with the increase in output.